P.J. Fogarty

On the exploration of innovative concepts for fusion chamber technology

Abstract This study, called APEX, is exploring novel concepts for fusion chamber technology that can substantially improve the attractiveness of fusion energy systems. The emphasis of the study is on fundamental understanding and advancing the underlying engineering sciences, integration of the physics and engineering requirements, and enhancing innovation for the chamber technology components surrounding the plasma. The chamber technology goals in APEX include: (1) high power density capability with neutron wall load >10 MW/m 2 and surface heat flux >2 MW/m 2 , (2) high power conversion efficiency (>40%), (3) high availability, and (4) simple technological and material constraints. Two classes of innovative concepts have emerged that offer great promise and deserve further research and development. The first class seeks to eliminate the solid “bare” first wall by flowing liquids facing the plasma. This liquid wall idea evolved during the APEX study into a number of concepts based on: (a) using liquid metals (Li or Sn–Li) or a molten salt (Flibe) as the working liquid, (b) utilizing electromagnetic, inertial and/or other types of forces to restrain the liquid against a backing wall and control the hydrodynamic flow configurations, and (c) employing a thin (∼2 cm) or thick (∼40 cm) liquid layer to remove the surface heat flux and attenuate the neutrons. These liquid wall concepts have some common features but also have widely different issues and merits. Some of the attractive features of liquid walls include the potential for: (1) high power density capability; (2) higher plasma β and stable physics regimes if liquid metals are used; (3) increased disruption survivability; (4) reduced volume of radioactive waste; (5) reduced radiation damage in structural materials; and (6) higher availability. Analyses show that not all of these potential advantages may be realized simultaneously in a single concept. However, the realization of only a subset of these advantages will result in remarkable progress toward attractive fusion energy systems. Of the many scientific and engineering issues for liquid walls, the most important are: (1) plasma–liquid interactions including both plasma–liquid surface and liquid wall–bulk plasma interactions; (2) hydrodynamic flow configuration control in complex geometries including penetrations; and (3) heat transfer at free surface and temperature control. The second class of concepts focuses on ideas for extending the capabilities, particularly the power density and operating temperature limits, of solid first walls. The most promising idea, called EVOLVE, is based on the use of a high-temperature refractory alloy (e.g. W–5% Re) with an innovative cooling scheme based on the use of the heat of vaporization of lithium. Calculations show that an evaporative system with Li at ∼1 200°C can remove the goal heat loads and result in a high power conversion efficiency. The vapor operating pressure is low, resulting in a very low operating stress in the structure. In addition, the lithium flow rate is about a factor of ten lower than that required for traditional self-cooled first wall/blanket concepts. Therefore, insulator coatings are not required. Key issues for EVOLVE include: (1) two-phase heat transfer and transport including MHD effects; (2) feasibility of fabricating entire blanket segments of W alloys; and (3) the effect of neutron irradiation on W.

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

Abstract Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W–5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. Systems study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kW h. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

Study of a spherical tokamak based volumetric neutron source

Abstract With the worldwide development of fusion power focusing on the design of the International Thermonuclear Experimental Reactor (ITER), developmental strategies for the demonstration fusion power plant (DEMO) are being discussed. A relatively prudent strategy is to construct and operate a small deuterium–tritium fuelled volumetric neutron source (VNS) in parallel with ITER. The VNS is to provide, over a period less than 20 years, a relatively high fusion neutron fluence of 6 MW year m−2 and wall loading of 1 MW m−2 or more, over an accessible blanket test area of more than 10 m2. Such a VNS would complement ITER in testing, developing, and qualifying nuclear technology components, materials, and their combinations for DEMO and future commercial power plants. The effort of this study has established the potential of the spherical tokamak as a credible VNS concept that satisfies the above requirements.

An overview of the iter in-vessel coil systems

ELM mitigation is of particular importance in ITER in order to prevent rapid erosion or melting of the divertor surface, with the consequent risk of water leaks, increased plasma impurity content and disruptivity. Exploitable “natural” small or no ELM regimes might yet be found which extrapolate to ITER but this cannot be depended upon. Resonant Magnetic Perturbation has been added to pellet pacing as a tool for ITER to mitigate ELMs. Both are required, since neither method is fully developed and much work remains to be done. In addition, in-vessel coils enable vertical stabilization and RWM control. For these reasons, in-vessel coils (IVCs) are being designed for ITER to provide control of Edge Localized Modes (ELMs) in addition to providing control of moderately unstable resistive wall modes (RWMs) and the vertical stability (VS) of the plasma.

Design of the national compact stellarator experiment (NCSX)

Abstract The National Compact Stellarator Experiment (NCSX) [ http://www.pppl.gov/ncsx/Meetings/CDR/CDRFinal/EngineeringOverview_R2.pdf ] is being designed as a proof of principal test of a quasi-axisymmetric compact stellarator. This concept combines the high beta and good confinement features of an advanced tokamak with the low current, disruption-free characteristics of a stellarator. NCSX has a three-field-period plasma configuration with an average major radius of 1.4 m, an average minor radius of 0.33 m and a toroidal magnetic field on axis of up to 2 T. The stellarator core is a complex assembly of four coil systems that surround the highly shaped plasma and vacuum vessel. Heating is provided by up to four, 1.5 MW neutral beam injectors and provision is made to add 6 MW of ICRH. The experiment will be built at the Princeton Plasma Physics Laboratory, with first plasma expected in 2007.

Free surface heat transfer and innovative designs for thin and thick liquid walls

Abstract Design windows on free surface flows in the APEX (advanced power extraction) study are derived from the viewpoints of the free surface heat transfer, the adaptation of liquid flows to the topological constraints, and temperature requirements for plasma operation and power conversion efficiency. Within these constraints, the temperature of the free liquid surface facing the plasma is the most critical parameter governing the amount of liquid that evaporates into the plasma chamber. Present analyses show that a 2 cm or a 40 cm thick lithium layer can be established throughout the ARIES-RS reactor using a velocity of 10 m s −1 while operating under the plasma compatible surface temperature. However, like solid metallic walls, the liquid lithium walls require the use of electrical insulators to overcome the MHD drag. As for Flibe free surface flows, the MHD effect caused by interaction with the mean flow is negligible, while a fairly uniform flow of 2 or 45 cm thick can be maintained throughout the reactor based on 3-D hydrodynamics calculations. However, being a low thermally conducting medium, the Flibe surface temperature highly depends on the extent of the turbulent convection. The heat transfer analyses based on the κ – e model of the turbulence, including MHD effects and various boundary conditions, predict a range of temperatures that may be beyond the plasma compatible temperatures. If indeed the Flibe surface temperature is high relative to the plasma operation limit, further design adjustments will be required to accommodate this deficiency.

Progress in NCSX Construction

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). Its mission is to develop the physics understanding of the compact stellarator and evaluate its potential for future fusion energy systems. Compact stellarators use 3D plasma shaping to produce a magnetic configuration that can be steady state without current drive or feedback control of instabilities. The NCSX has major radius 1.4 m, aspect ratio 4.4, 3 field periods, and a quasi-axisymmetric magnetic field. It is predicted to be stable and have good magnetic surfaces at beta > 4% and to have tokamak-like confinement properties. The device will provide the plasma configuration flexibility and the heating and diagnostic access needed to test physics predictions. Component production has advanced substantially since the first contracts were placed in 2004. Manufacture of the vacuum vessel was completed in 2006. All eighteen modular coil winding forms have been delivered, and twelve modular coils have been wound and epoxy impregnated. A contract for the (planar) toroidal field coils was placed in 2006 and manufacture is in progress. Assembly activities have begun and will be the projects main focus in the next few years. The engineering challenge of NCSX is to meet the requirements for complex geometries and tight tolerances within the cost and schedule constraints of a construction project. This paper will focus on how the engineering challenges of component production have been resolved, and how the assembly challenges are being met.

Design of the quasi-poloidal stellarator experiment (QPS)

Abstract The Quasi-poloidal stellarator (QPS), currently in the early design phase, is a low-aspect-ratio (R/a=2.7), concept exploration experiment with a non-axisymmetric, near-poloidally-symmetric magnetic configuration. The QPS design parameters are 〈R〉=0.9 m, 〈a〉=0.33 m, B=1 T, and a 1 s pulse length. The QPS device will be located at the Oak Ridge National Laboratory. Lyon et al. [ http://qps.fed.ornl.gov/pvr/pdf/qpsentire.pdf , 2001] describes the physics and engineering features in detail. The QPS device is estimated to require 4 years from start of design to first plasma in 2007.

Conceptual design of the vacuum pumping system for the Tokamak Physics Experiment

The conceptual design of the TPX vacuum pumping system is presented. The baseline concept includes a high vacuum pumping system, a roughing and backing system, volume pumping ducts, a leak detection system, a diagnostic pumping system, and a cryostat pumping system. The high vacuum pumping system will initially evacuate the torus, provide pumping of the diverters during operation, and provide pumping for glow discharge cleaning. The high vacuum pumping system has high throughput and variable conductance capabilities and includes cryocondensation pumps for pumping deuterium during normal operation as well as turbomolecular pumps for pumping helium and for glow discharge cleaning. Sixteen vacuum ducts extend from the vacuum vessel through the cryostat to the pumping system; each duct contains a torus isolation valve and an electrical break. Butterfly valves at the cryopump inlets will be used for throttling the pumps and for pump regeneration. In this way, half of the pumps can be regenerated while the others are operating. The specific design parameters and predicted performance of the vacuum pumping system are discussed, as are the upgrade options for steady state and DT operation.

Lessons learned during the manufacture of the NCSX modular coils

The National Compact Stellarator Experiments (NCSX) modular coils presented a number of engineering and manufacturing challenges due to their complex shapes, requirements for high dimensional accuracy and high current density requirements due to space constraints. Being the first of their kind, these coils required the implementation of many new manufacturing and measuring techniques and procedures. This was the first time that these manufacturing techniques and methods were applied in the production of coils at the laboratory. This resulted in a steep learning curve for the first several coils. Through the effective use of procedures, tooling modifications, involvement and ownership by the manufacturing workforce, and an emphasis on safety, the assembly team was able to reduce the manufacturing times and improve upon the manufacturing methods. This paper will discuss the learning curve and steps that were taken to improve the manufacturing efficiency and reduce the manufacturing times for the modular coils without forfeiting quality.